• Proceedings of the KSME Conference
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• 2004.04a
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• pp.78-83
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• 2004

Environmental fatigue crack propagation of CF8M and CF8A steels used in the domestic PWR were investigated on the simulated PWR condition(Temperature: $316^{\circ}C$, Pressure: 15MPa). The test equipment for environmental fatigue(high temperature-high pressure loop, autoclave, load frame, measurement system) were designed. As-received and 60-year aged specimens were used in the test. To compare with environmental fatigue test, another test was performed in the air condition. The fracture surface of specimens were difficult to verify the fracture modes such as striation, intergranular crack and cleavage and so on. As the ferrite content of CF8M is increased, more particles covered fracture surface were peeled.

• Korean Journal of Materials Research
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• v.25 no.4
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• pp.183-190
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• 2015

Fatigue crack growth rate tests were conducted as a function of temperature, dissolved hydrogen (DH) level, and frequency in a simulated PWR environment. Fatigue crack growth rates increased slightly with increasing temperature in air. However, the fatigue crack growth rate did not change with increasing temperature in PWR water conditions. The DH levels did not affect the measured crack growth rate under the given test conditions. At $316^{\circ}C$, oxides were observed on the fatigue crack surface, where the size of the oxide particles was about $0.2{\mu}m$ at 5 ppb. Fatigue crack growth rate increased slightly with decreasing frequency within the frequency range of 0.1 Hz and 10 Hz in PWR water conditions; however, crack growth rate increased considerably at 0.01 Hz. The decrease of the fatigue crack growth rate in PWR water condition is attributed to crack closure resulting from the formation of oxides near the crack tips at a rather fast loading frequency of 10 Hz.

• Proceedings of the KSME Conference
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• 2004.11a
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• pp.178-183
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• 2004

For developing fatigue design curve of cast stainless steels that would be used in piping material of domestic nuclear power plants, a low-cycle fatigue test rig was built. It is capable of performing tests in pressurized high temperature water environment of PWR. Cylindrical specimens of CF8M were used for the strain-controlled environmental fatigue tests. Fatigue life was measured in terms of the number of cycles with the variation of strain amplitude at 0.04%/s strain rates. The disparity between target length and measured length of specimens was corrected by using finite element method. The corrected test results showed similar fatigue life trend with another previous results.

• Journal of Radiation Protection and Research
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• v.42 no.2
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• pp.106-113
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• 2017

Background: Interfacing-system loss-of-coolant-accident (ISLOCA) has been identified as the most hazardous accident scenario in the typical PWR plants. The present study as an effort to improve the knowledge of the source term to the environment during ISLOCA focuses on an improvement of the estimation method. Materials and Methods: The improvement was performed to take into account an effect of broken pipeline and auxiliary building structures relevant to ISLOCA. An estimation of the source term to the environment was for the OPR-1000 plants by MELOCR code version 1.8.6. Results and Discussion: The key features of the source term showed that the massive amount of fission products departed from the beginning of core degradation to the vessel breach. Conclusion: The release amount of fission products may be affected by the broken pipeline and the auxiliary building structure associated with release pathway.

• Proceedings of the Materials Research Society of Korea Conference
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• 2011.10a
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• pp.39.1-39.1
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• 2011

가압경수로의 압력경계기기는 약 $300^{\circ}C$, 150기압의 고온고압수환경에서 가동되고 있다. 특히 가압기 밀림관은 고온수와, 저온수가 교차하는 부분으로 열성층 형성으로 열적, 기계적 피로 및 수화학환경이 더해진 부식피로 등에 의하여 손상을 받는다. PWR 원전에서 수화학환경은 대표적으로 용존산소(DO) 5ppb, pH 6~8, 용존수소(DH) <30 cc/kg, 온도 $316^{\circ}C$의 환경을 유지하게 된다. 가압기 밀림관에는 오스테나이트계 스테인리스강이 사용되는데, 오스테나이트계 스테인리스강은 고온 수화학환경에 민감한 것으로 알려져 있다. 따라서 오스테나이트계 스테인리강을 공기중에서의 기계적특성 및 피로특성을 향상시키기 위하여 질소를 첨가한 스테인리스강을 제조하여 PWR 원전환경에서의 피로균열성장특성을 평가하였다. 실험에 사용된 재료는 PWR 원전 가압기 밀림관 소재인 Type 347 스테인리스강에 0.0005 wt%가 첨가된 상용재와 0.11 wt% 질소가 첨가된 재료이다. 사용된 시편형상은 두께 5 mm, 폭 25.4 mm의 CT 시편이다. 수화학환경은 150기압, 온도 $316^{\circ}C$, 용존산소(DO) 5ppb, 용존수소(DH) 30 cc/Kg, pH는 약 7로 유지 하였으며, 응력비 0.1, 하중 반복속도 10Hz의 기계적 조건에서 하중제어로 시험을 진행하였다. 균열길이는 직류전위차법(Direct Current Potential Drop: DCPD)을 이용하여 측정하였다. 질소함량이 증가할수록 동일 사이클에서 균열길이가 늦게 성장하였고, 피로균열성장속도도 약간 늦어지는 것으로 나타났다. 각 스테인리스강의 피로파면 관찰결과 상용재는 약 1 ${\mu}m$의 산화물들이 생성되는 반면 질소첨가 스테인리스강은 약 0.1 ${\mu}m$정도 산화물이 생성되었다. 산화막의 두께도 질소가 첨가됨으로써 상용재에 비해 얇게 생성되었다. 따라서 질소가 첨가됨으로써 부식환경에서 내산화성이 향상되었으며, 이는 피로균열성장특성에 영향을 미치는 것으로 판단된다.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.307-308
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• 2004

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.621-626
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• 2003

A DUPIC nuclear fuel is a newly developed fuel for CANDU reactors based on the concept of refabrication of spent PWR fuel by a dry process. Because a spent PWR fuel, a highly radioactive material, is used as a starting material, the experimental verification of DUPIC nuclear fuel fabrication requires an appropriate facility which should satisfy engineering requirements and guarantees safe operation. DUPIC nuclear fuel development team modified M6 hot-cell in IMEF to construct the dedicated facility(DFDF) for tile experiment. The experiment with spent PWR fuel have been conducted since January of 2000. Environmental effects of DFDF normal operation have been investigated when DUPIC nuclear fuel is fabricated with the maximum capacity of 50kg U/yr. The analysis results of the radiological safety of DFDF facility have shown that both national regulation limit and IMEF design criteria are satisfied.

• Nuclear Engineering and Technology
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• v.49 no.4
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• pp.752-759
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• 2017

The fatigue crack growth behavior of Type 347 (S347) and Type 347N (S347N) stainless steel was evaluated under the operating conditions of a pressurized water reactor (PWR). These two materials showed different fatigue crack growth rates (FCGRs) according to the changes in dissolved oxygen content and frequency. Under the simulated PWR conditions for normal operation, the FCGR of S347N was lower than that of S347 and insensitive to the changes in PWR water conditions. The higher yield strength and better corrosion resistance of the nitrogen-alloyed Type 347 stainless steel might be a main cause of slower FCGR and more stable properties against changes in environmental conditions.

• Nuclear Engineering and Technology
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• v.46 no.3
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• pp.431-438
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• 2014

The effect of various pH agents on the corrosion behavior of carbon steel was investigated under a simulated secondary water chemistry condition of a pressurized water reactor (PWR) in a laboratory, and the steel's corrosion performance was compared with the field data obtained from Uljin NPP unit 2 reactor. All tests were carried out at temperatures of $50^{\circ}C-250^{\circ}C$and pH of 8.5 - 10. The pH at a given temperature was controlled by adding different agents. Laboratory data indicate that the corrosion rate of carbon steel decreased as the pH increased under the test conditions and the highest corrosion rate was measured at $150^{\circ}C$. This high corrosion rate may be related to high dissolution and instability of Fe oxide ($Fe_3O_4$) at $150^{\circ}C$. It was also found that an addition of ethanolamine (ETA) to ammonia was more effectivefor anticorrosion than ammonia alone, and that mixed treatment reduced 50% of iron or more at pHs of 9.5 or higher, especially in the steam generator (SG) and the moisture separator & re-heater (MSR).

• Korean Journal of Metals and Materials
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• v.47 no.12
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• pp.819-827
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• 2009

This work is concerned with a statistical analysis of factors affecting the irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steels for core internals of pressurized water reactors (PWR). The microstructural and environmental factors were reviewed and critically evaluated by the statistical analysis. The Cr depletion at grain boundary was determined to have no significant correlation with the IASCC susceptibility. The threshold irradiation fluence of IASCC in a PWR was statistically calculated to decrease from 5.799 to 1.914 DPA with increase of temperature from 320 to $340^{\circ}C$. From the analysis of the relationship between applied stress and time-to-failure of stainless steel components based on an accelerated life testing model, it was found that B2 life of a baffle former bolt exposed to neutron fluence of 20 and 75 DPA was at least 2.5 and 0.4 year, respectively, within 95% confidence interval.